First Wall Research Articles

Depth-resolved deuterium retention analysis in displacement-damaged tungsten using laser-induced breakdown spectroscopy

Fuel retention in plasma-facing components (PFCs) is a critical issue in future nuclear fusion reactors operating with Deuterium-Tritium (DT) regarding nuclear safety and fulfillment of the T cycle. However, during DT plasma operation, highly energetic neutrons will induce damage in the lattice of W PFCs causing enhanced fuel retention in defects or traps. Laser-Induced Breakdown Spectroscopy (LIBS) is a potential tool to monitor the T-content in situ in PFCs of future nuclear fusion devices. This article presents an ex situ study on pre-damaged W material after D plasma exposure to qualify the method and mimic conditions expected in a reactor. ITER grade W samples were displacement-damaged by 10.8 MeV W ions to a damage dose of 0.23 dpa and exposed to low temperature deuterium plasma at low energy in PlaQ. The resulting deuterium concentration was analyzed by using 3He Nuclear Reaction Analysis (depth resolution of ≈150 nm) as a well-established method, and LIBS (picosecond laser pulses, depth resolution of 15 nm). The sample with the highest deuterium concentration showed a deuterium-rich zone up to a depth of 1.13 μm using both techniques. This is close to the expected W ion-induced damage depth of ≈1 μm. The results imply that LIBS as an in situ technique for tritium monitoring could be a viable option for a reactor.

Microstructure of additive manufactured materials for plasma-facing components of future fusion reactors

Microstructure of additive manufactured materials for plasma-facing components of future fusion reactors

A novel plasma-facing NdB6 particulate reinforced W1Ni matrix composite: Powder metallurgical fabrication, microstructural and mechanical characterization

Tungsten (W) is one of best candidate metal for plasma-facing materials (PFM), especially due to its high melting temperature and neutron absorption capability. However, converting W into bulk PFM is hard because of its high melting point. This problem can be solved by adding metallic sintering aids with low melting points. In this study, W matrix with 1 wt% Ni aid was reinforced by adding NdB6 particles (1, 5, and 10 wt%). It can be introduced as a novel potential PFM, thanks to its low volatility and high neutron absorbability. The ceramic and composite powders produced via mechanochemical synthesis and mechanical alloying were examined in terms of composition, particle size, crystallite size, and lattice strain. Samples sintered via pressureless sintering (PS) and spark plasma sintering (SPS) were microstructurally analyzed by using an X-ray diffractometer (XRD), a scanning electron microscope (SEM) attached with an energy dispersive spectroscope (EDS), and mechanically analyzed in terms of microhardness and wear behavior. Based on the results, W2B and WB phases emerged in the SPS'ed W1Ni-5NdB6 and PS'ed./SPS'ed W1Ni-10NdB6 composites. SPS'ed W1Ni-10NdB6 composite had the highest hardness value and the lowest specific wear rate. The SPS'ed W1Ni-5NdB6 composite showed fewer surface damages and higher irradiation resistance as compared with other samples after exposure of He+ irradiation.

Synergistical improvement in toughness and strength of multi-doped W-Hf-Y2O3 alloy via formation of complex oxide particles

As the primary candidate material for plasma facing components (PFCs) in fusion reactor, the serious brittleness of tungsten has become an urgent issue to be solved. In this work, a multi-doped W-Hf-Y2O3 alloy was prepared by high-energy ball milling and spark plasma sintering, aiming at improving its toughness through formation of complex oxide particles. Complex Y2Hf2O7 with pyrochlore structure were developed in the multi-doped WHY alloy due to dissolution and precipitation mechanism, which was confirmed by the combined characterization results of X-ray diffraction (XRD), electron probe microanalysis (EPMA) and transmission electron microscope (TEM). Complex oxide particles of Y2Hf2O7 with average size of 38.8 nm were densely distributed in the alloy. The formation of complex Y2Hf2O7 particles with nano-scale size not only consumed the impurity of oxygen, but also played a critical role in grain refinement because of Zener pinning effect. Improvement in mechanical properties of the multi-doped WHY alloy was then achieved, and the fracture toughness and compressive strength reached 19.7 MPa·m1/2 and 2223.4 MPa, respectively.

Helium plasma operations on ASDEX Upgrade and JET in support of the non-nuclear phases of ITER

For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of the EUROfusion Consortium carried out dedicated experimental campaigns in He on the ASDEX Upgrade (AUG) and JET tokamaks in 2022, with particular emphasis put on the ELMy H-mode operation and plasma-wall interaction processes as well as comparison to H or deuterium (D) plasmas. Both in pure He and mixed He + H plasmas, H-mode operation could be reached but more effort was needed to obtain a stable plasma scenario than in H or D. Even if the power threshold for the LH transition was lower in He, entering the type-I ELMy regime appeared to require equally much or even more heating power than in H. Suppression of ELMs by resonant magnetic perturbations was studied on AUG but was only possible in plasmas with a He content below 19%; the reason for this unexpected behaviour remains still unclear and various theoretical approaches are being pursued to properly understand the physics behind ELM suppression. The erosion rates of tungsten (W) plasma-facing components were an order of magnitude larger than what has been reported in hydrogenic plasmas, which can be attributed to the prominent role of He2+ ions in the plasma. For the first time, the formation of nanoscale structures (W fuzz) was unambiguously demonstrated in H-mode He plasmas on AUG. However, no direct evidence of fuzz creation on JET was obtained despite the main conditions for its occurrence being met. The reason could be a delicate balance between W erosion by ELMs, competition between the growth and annealing of the fuzz, and coverage of the surface with co-deposits.

Development and manufacturing of beryllium-armoring ITER enhanced heat flux FW toward series production in China

ITER’s enhanced heat flux (EHF) first wall (FW) provides thermal shielding to the other components behind it and limits its plasma boundary, consequently bearing high surface heat loads up to 4.7 MW m−2. China developed the EHF FW technologies in steady stages by making small mock-ups, semi-prototypes and a full-scale prototype (FSP), working toward series production at the Southwestern Institute of Physics. All the key technologies for bimetallic diffusion bonding, welding and assembly have been fully qualified. The Be armor tile size effects on thermal-fatigue performance have been evaluated using a pulsed high heat flux test with an EMS-400 electron beam facility. It was found that both the Be/CuCrZr bonding and its thermal-fatigue life are highly dependent on the tile size, and 12 × 12 mm2 Be tiles are acceptable for the required performance, either as individual tiles or in castellation by cutting larger ones. Small-scale mock-ups with such Be tiles survived 16 000 thermal cycles at 4.7 MW m−2, while EHF FW fingers with 16 × 16 mm2 Be tiles demonstrated unstable performance. In contrast, only BE tiles larger than 24 × 12 mm2 showed reliable diffusion bonding with the CuCrZr heat sink by hot isostatic pressing and consequently should be castellated into smaller ones. The manufacturing of the FSP finger goes through complex thermal cycles, using CuCrZr alloys in age-strengthening plus cold-rolling states, enabling it to maintain its tensile strength to the level of 285 MPa and its mean grain size less than 100 µm. The pipe welding for finger pairing and its assembly with the central beam are demonstrated, including enabling welding to be carried out twice for repair in a narrow space. The FSP showed good finger alignment, dimensional control and reliable vacuum tightness without any blockage of cooling channels.

Sputtering yield increase with fluence in low-energy argon plasma-tungsten interaction

The increase of tungsten sputtering yield with fluence was investigated during plasma (≤100 eV) irradiation. This analysis focused on the combined effects of surface binding energy and surface morphology caused by Ar retention. As post-mortem analysis, Ar concentration was measured with SIMS (Secondary Ion Mass Spectroscopy) and TDS (Thermal Desorption Spectroscopy). The Ar concentration saturated at 21 % of W number density during the sputtering process. The corresponding change in surface morphology causes a change in the local ion incident angle, which leads to a sputtering yield increase of 10 %. The increased Ar concentration leads to a decrease in surface binding energy and a change in surface morphology which increases W sputtering yield from 0.02 to 0.03 by ion energy 80 eV. Over time, W sputtering yield reaches saturation as a function of saturated Ar concentration. This result implies that the synergistic role of Ar concentration and surface morphology on sputtering yield. This sputtering yield enhancement occurs more seriously in the realistic condition of plasma-facing materials that face low-energy plasma.

Divertor heat load estimates on NSTX and DIII-D using new and open-source 2D inversion analysis code

A thermography inversion algorithm has been developed in the open-source Python-based computer code, HYPERION, to calculate the heat flux incident on plasma-facing components (PFCs) in axisymmetric tokamaks. The chosen mesh size at the surface significantly affects the calculated transient heat flux results. The calculated transient heat flux will exceed the real value when the mesh size tends to zero but will underestimate the real value when the mesh size is large. A criterion for determining the appropriate mesh size for the transient heat flux calculation will be discussed. The numerical scheme for HYPERION uses a 2D fully implicit finite-difference approach, allowing temperature-dependent thermal properties of PFC materials. The inversion algorithm is benchmarked against established heat flux calculation codes, TACO and THEODOR, based on thermography data from NSTX and DIII-D respectively. The primary benefits of HYPERION compared to TACO and THEODOR are that it is open-source and it allows for the optimization of mesh thickness along the substrate. The algorithm also accounts for the thermal properties of thin surface layers that characteristically form on PFCs due to plasma-material interactions. The agreement between HYPERION and THEODOR is excellent, as the percent difference between the codes is ∼5% on average in the case of the DIII-D data for moderate to high heat flux. Verification tests with TACO show slightly higher average percent differences of 8% and 12%. In using HYPERION to study filaments in heat flux, the initial results indicate that small ELMs filaments significantly broaden the divertor heat flux, and decrease divertor peak flux. Compared to the inter-ELM, the small ELM filaments decrease the divertor peak surface temperature. With intermittent divertor filaments, the divertor heat flux width is comparable with that found in L-mode.

Assessment of the runaway electron load distribution in ITER during 3D MHD induced beam termination

In ITER, disruption-born runaway electrons (REs), unless mitigated, are expected to form a several Mega-Ampere beam that ultimately intercepts the first wall leading to melting of the plasma facing components. Developing a successful mitigation strategy therefore requires modeling that takes into account the coupling between REs and the MHD during the formation and termination of the beam, along with estimates for the wall loads that can be compared to design values. Using the JOREK code, this work aims to provide the latter by presenting a novel model for collisions between REs and the wall and applying it to a beam termination scenario in ITER. To this end, the transport of REs is modeled by tracing particles in the fields calculated by preceding simulations of the disruption event where REs were treated in the fluid picture and self-consistently coupled to the MHD. The resulting heat loads are found to be highly localized in both the poloidal and toroidal directions, with 3D features also playing an important role in the appearance of hot spots. Peak loads were on the order of 107−108Jm2 . The load distribution was found to only be weakly sensitive to the initial phase space distribution of the REs when decoupled from the fields, while modifying properties of the underlying MHD yielded notable differences in wetted area and peak loads, in particular for cases with higher resistivity.

Sputtering fabrication of isolated W nanocolumns: A possible alternative as plasma facing material for nuclear fusion reactors

Isolated tungsten nanocolumns (W-NCs) have been reported to exhibit a higher radiation resistance than coarse grained W samples, under radiation conditions similar to those that plasma facing materials would face in both magnetic and inertial confinement nuclear fusion approaches (MCF and ICF, respectively). This is so, because their ability to release He via the free surfaces and, the strong reduction of the sputtering yield under KeV Ar and D irradiation as well as, flattening of its angular dependence. The latter is very important for the divertor location in MCF.In this work, we investigate the capabilities of sputtering (an easy control, environmentally friendly, versatile, scalable and low-cost technique) to fabricate isolated W nanocolumns. We study the influence of sputtering parameters (plasma power and deposition angle) on the morphology, density and microstructure of the deposited coatings. Results show that stable α-phase 3D isolated W-NCs are produced at low plasma power (<100 W) and high deposition angles (θ ≥ 75°).

Non-linear MHD investigations of high-confinement regimes without type-I ELMs in ASDEX Upgrade and JT-60SA

Large edge localised modes (ELMs) would cause an unacceptable reduction of material lifetime in future large tokamaks due to the significant amount of energy expelled from the magnetically confined region towards the plasma facing components. Thoroughly validated modelling of regimes devoid of large ELMs is crucial as it may then provide predictive insights prior to tokamak operation and design. This paper describes recent efforts pursued with the non-linear extended MHD code JOREK in the modelling of three scenarios without large ELMs: quiescent H-mode (QH-mode), quasi-continuous exhaust regime (QCE regime), and the enhanced D-alpha H-mode (EDA H-mode). For each of these regimes, the non-linear dynamics observed in the simulations are detailed and compared to experimental observations of the underlying instabilities of each regime (edge harmonic oscillation for QH-mode, small ELMs for QCE regime, and quasi-coherent mode for EDA H-mode). For QH-mode, the kink-peeling mode is found to govern the dynamics and a transition to a large ELM is obtained above the same density threshold as in the modelled experiment. For the QCE regime and EDA H-mode, resistive peeling–ballooning modes dominate and pedestal fluctuation frequencies correspond well to experimental observations. The dominant mechanisms for the excitation and suppression of these instabilities are presented and their influence on simulation dynamics is shown. Finally, predictive simulations of edge instabilities at different values of plasma resistivity in a 4.60 MA scenario with low edge safety factor in JT-60SA are presented.

Behavior of Foreign Interstitial Hydrogen and Helium Atoms at Tungsten/Beryllium Interface from First-Principles Calculations

The behavior of foreign interstitial hydrogen and helium atoms and its effect on the physical properties of the tungsten/beryllium interface structure were computationally studied by first-principles calculations. Briefly, as part of our study of helium irradiation damage and hydrogen detention, the following properties were calculated: (1) the electronic properties of the tungsten/beryllium interface structure with a single interstitial hydrogen or helium atom and Hen vacancy or Hn vacancy complexes, and (2) the isotropy (polycrystalline) elastic modulus (bulk, torsion, Young’s modulus), anisotropy factor and minimum thermal conductivity of the previously described tungsten/beryllium interface systems. This study found that defect atoms are more likely to be concentrated in beryllium, but the tungsten layer is more sensitive to changes in mechanical properties caused by interstitial atoms. The ability of the beryllium vacancies to capture interstitial atoms is smaller than that of the tungsten vacancies. Based on the computational results, a preliminary assumption of the judgment of the tungsten/beryllium interface structure on the resistivity for plasma-facing materials is introduced. These computational studies provide a critical evaluation of the radiation resistivity and hydrogen retention of tungsten/beryllium interface materials. The calculated interface properties can be incorporated into radiation damage resistance property evaluation systems to develop and test tungsten-based composite materials.

Design of a 3D-printed liquid lithium divertor target plate and its interaction with high-density plasma

A liquid Li divertor is a promising alternative for future fusion devices. In this work a new divertor model is proposed, which is processed by 3D-printing technology to accurately control the size of the internal capillary structure. At a steady-state heat load of 10 MW m−2, the thermal stress of the tungsten target is within the bearing range of tungsten by finite-element simulation. In order to evaluate the wicking ability of the capillary structure, the wicking process at 600 °C was simulated by FLUENT. The result was identical to that of the corresponding experiments. Within 1 s, liquid lithium was wicked to the target surface by the capillary structure of the target and quickly spread on the target surface. During the wicking process, the average wicking mass rate of lithium should reach 0.062 g s−1, which could even supplement the evaporation requirement of liquid lithium under an environment > 950 °C. Irradiation experiments under different plasma discharge currents were carried out in a linear plasma device (SCU-PSI), and the evolution of the vapor cloud during plasma irradiation was analyzed. It was found that the target temperature tends to plateau despite the gradually increased input current, indicating that the vapor shielding effect is gradually enhanced. The irradiation experiment also confirmed that the 3D-printed tungsten structure has better heat consumption performance than a tungsten mesh structure or multichannel structure. These results reveal the application potential and feasibility of a 3D-printed porous capillary structure in plasma-facing components and provide a reference for further liquid−solid combined target designs.

Manufacturing trials of PFCs with low thermal conductivity features for limiters

In future demonstration fusion power plants, plasma facing components will face high steady state and ultra-high transient heat loads from the plasma. Introducing low thermal conductivity features to the component can reduce the peak temperatures seen by the coolant pipe during transient loads, but at the trade-off of a loss of steady-state performance. Producing low-conductivity tungsten by additive methods has been investigated elsewhere, so this trial investigates production through subtractive means i.e., conventional milling. This method preserves the original temperature limits and majority of the microstructure and is much higher Technology Readiness Level (TRL) so requires less development to reach series manufacture. In this trial, diamond drilling produced holes in monoblocks down to 0.6 mm and ligaments down to 0.7 mm, showing that this method is capable of producing small features in a controlled way. With the geometric design of lattice used in this trial, these samples achieved thermal conductivity reductions of 15% to 40%. One monoblock assembly was tested up to 1100 °C for 500 cycles in the Heating by Induction to Verify Extremes (HIVE) High Heat Flux test rig and showed no signs of ligament cracking or thermal degradation. Electromagnetic modelling has been validated against the experimental results, so future designs can be accurately modelled ahead of testing in HIVE.

Interfaces enhanced plasma irradiation resistance in CrMoTaWV/W multilayer films through blocking He diffusion

The performance of plasma-facing materials (PFMs) is one of the key factors that significantly impact the stability of operation in fusion reactors. Herein, a new CrMoTaWV/W (high entropy alloy (HEA)/W) multilayer structure is designed as PFM to investigate its resistance to He plasma irradiation. It was observed that the introduction of the interfaces effectively absorbed plenty of He atoms, preventing them from diffusing into the material and delaying the formation of fuzz incubation zone, therefore, enhancing the resistance to plasma irradiation. The thickness transformed to fuzz in the HEA/W multilayer films was observed to be about two-thirds of those in the CrMoTaWV (HEA) film. Additionally, the fuzz growth rates in HEA/W multilayer films are lower than the average growth rate of bulk W and HEA films combined. These findings highlight a promising new avenue for the exploration of high-performance PFMs.

Tungsten wall cratering under high-velocity dust impacts: Influence of impact angle and temperature

The integrity of plasma-facing components (PFCs) in fusion reactors is severely tested by high-velocity dust collisions, which occur during explosive events such as runaway electron terminations. These events can expel dust particles at velocities of 0.5 – 1 km/s in current fusion devices and potentially several km/s in advanced reactors like ITER and DEMO, leading to significant material erosion and damage. Given the limitations of existing models, which effectively address only low-velocity impacts, there is a critical need for improved modeling of high-velocity dust-wall interactions. This study utilizes molecular dynamics (MD) simulations to explore the effects of impact angle and target temperature on the interactions between tungsten (W) dust particles and W walls under extreme velocities ranging from 2.5 to 4.5 km/s. Our research focuses on analyzing the morphology of impact craters, and characteristics of ejecta across a range of impact angles (0° to 75°) and with dislocation density for temperatures (300 to 3000 K). Our study reveals that the angle of impact and temperature almost exclusively determine the shape of the crater and the distribution of ejecta, highlighting the critical role of these factors in the dynamics of dust-wall interactions. Comparison with the experimental data obtained from W-on-W impact tests shows a strong correlation with our theoretical predictions.

Improvement in thermal stability of ODS-W alloy through formation of complex oxide dispersoids

Thermal stability is a crucial issue for engineering application of tungsten, especially as plasma facing material in future fusion reactor. Two kinds of ODS-W alloys of W-0.5Y2O3 (WY) and W-0.5Y2O3-0.25Ti (WYT) were fabricated by ball milling and spark plasma sintering aiming at formation of oxide dispersoids with different crystallographic structures. Thermal stability of both ODS-W alloys was evaluated in terms of microhardness and microstructure evolution during isothermal aging at 1100 °C. Dispersoids of sesquioxide Y2O3 were identified in sintered WY, while pyrochlore Y2Ti2O7 with finer size and higher number density were developed in sintered WYT, resulting in higher bending strength of 1020 MPa and microhardness of 880 HV. Compared with sesquioxide Y2O3, complex dispersoids of pyrochlore Y2Ti2O7 exhibited much better coarsening resistance during isothermal aging at 1100 °C up to 100 h, where the intracrystalline and GB dispersoids maintained around 10 nm and 30 nm, respectively. The excellent coarsening resistance of complex Y2Ti2O7 with pyrochlore structure contributed to grain refinement and improved microstructure stability of WYT due to the effect of Zener drag.

DBTT and recrystallization behavior analyses for rolled and forged potassium-doped tungsten alloys

Tungsten-potassium (WK) alloys are regarded as one of the candidates for plasma-facing materials (PFMs) intended for the divertor and first wall of future fusion reactors. Determining the range of the ductile-to-brittle transition temperature (DBTT) and the recrystallization temperature (RCT) for the WK alloy is critical as it is closely related to the safe operational temperature window of the fusion reactor. Variable temperature tensile tests were conducted on rolled and high energy rate forging (HERF) WK alloy specimens to elucidate their DBTT. Additionally, gradient high-temperature isochronal annealing was performed on the HERF samples, coupled with Vickers hardness characterization, to ascertain their distinct RCT for various orientation planes within the same alloy. Subsequently, the DBTT of the as-rolled WK alloy is determined to be approximately between 100 °C and 150 °C, while that of the HERF WK alloy is found to be around lower than 200 °C. Furthermore, the RCT of the ND plane of the HERF WK alloy is approximately 1600 °C, whereas that of the RD counterpart exceeded 1700 °C. The internal mechanisms underlying these phenomena are comprehensively analyzed in this study. This work has identified the safe service temperature window for future fusion reactors with this kind of advanced WK alloy and provided a new perspective for determining the RCT of many other W-based materials.

Study on hydrogen retention resistance of diamond film by HFCVD

To explore the potential application of diamond coating on the surface modification of graphite-based plasma-facing components in nuclear fusion, the retention of hydrogen irradiated into diamond film deposited by HFCVD was studied. After being attacked with 200 eV hydrogen ions for 10 h at 200 °C, this diamond layer was observed to maintain the stability of crystal structure. The C-H absorption peaks in the FTIR spectrum that were conspicuous in the sole graphite were absent in the diamond surface subjected to the same quantity of hydrogen irradiation, suggesting that the implanted hydrogen was difficult to penetrate or persist in the diamond covering. Further computational studies have demonstrated that hydrogen atoms exhibit a low propensity for retention within the diamond (111), in contrast to reside stably in graphite or graphite-like structures. The above shows that HFCVD diamond film has strong resistance to hydrogen retention and structural stability under hydrogen irradiation.

Comprehensive new insights on the potential use of SiC as plasma-facing materials in future fusion reactors

The performance of silicon carbide as an alternative plasma facing material (PFM) was studied at various irradiation conditions relevant to ion energies and fluxes of a fusion reactor. This analysis involves detailed modeling of subsurface plasma/material interactions, sputtered particle transport above the surface and redeposition, and related changes in material composition and microstructure induced by steady-state and Edge Localized Mode ion fluxes. Transition of a crystalline SiC surface to semi-crystalline and amorphous phases was analyzed based on advanced modeling of DIII-D tokamak experiments where SiC was irradiated in single- and multiple- L-mode and H-mode discharges. This analysis shows that displacement damage, particle deposition/redeposition, and D accumulation on the SiC divertor surface can lead to significant microstructural changes that result in enhanced sputtering erosion in comparison with the original crystalline material. However, the resulting total net erosion rate for a full-coverage, advanced tokamak, SiC coated divertor may well be acceptably low. Moreover, the C sputtering yield from the evolved SiC surface can be seven times lower than from a pure graphite surface; this would imply significantly reduced tritium co-deposition rates in a D-T tokamak reactor, compared with a pure carbon surface. It was also determined that chemical sputtering of both C and Si should not result in any noticeable effect on the net erosion, for attached plasma regimes. Our results thus show encouraging results overall for use of SiC as a PFM in tokamaks.